1. Field of the Invention
The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, but not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.
2. Description of the Related Art
At the present time more than 50% of the world's annual production of radionuclides is used for medical purposes, such as for the early diagnoses of diseases and for therapy. A basic condition of the use of radionuclides in medicine is the requirement that the radiation exposure to a patient be minimal. This necessitates the use of short-lived radionuclides. A nuclide with a short half-life, however, creates difficulties in transportation and storage. One of the radionuclides used most often for medical purposes is molybdenum-99 with a half-life of 66 hours. Molybdenum-99 decay results in technetium-99m with a half-life of 6 hours and gamma energy at 140 keV, which is convenient for detection. Currently, more than 70% of diagnostic examinations are performed using this radionuclide.
One method of molybdenum-99 production involves using a target of natural molybdenum or molybdenum enriched in molybdenum-98 irradiated by a neutron flux in a nuclear reactor. Molybdenum-99 results from a neutron radiation capture by 98Mo(n,γ)99Mo. The irradiated target containing molybdenum-99 then undergoes chemical purification. This method, however, has a low yield and the molybdenum-99 produced is characterized by a low specific activity due to the presence of unreacted molybdenum-98 in the final product.
Another method of molybdenum-99 production is based on uranium fission under neutron irradiation of a U—Al alloy or electroplated target in a nuclear reactor. The target contains high enriched uranium (HEU) which typically contains greater than about 85 weight percent uranium-235, which is also considered weapons grade uranium. After irradiation, the target is processed by one of the traditional chemical methods to extract molybdenum-99 from the fission products. The specific activity of molybdenum-99 achieved by this method is several tens of kilocuries per gram of molybdenum. A serious disadvantage of this method is the production and disposition of large amounts of radioactive wastes that are byproducts of the fission process, including some un-fissioned uranium. These targets are single-use. The activity of these wastes exceeds that of the molybdenum-99 material produced by two orders of magnitude. A twenty-four hour delay in processing the irradiated uranium targets results in a decrease of total activity by about an order of magnitude, during which time the molybdenum-99 activity decreases by approximately 22%. After two days, the activity of the waste byproducts exceeds that of the molybdenum-99 by a factor of six to seven. The problem of long-lived fission product management and security of the residual HEU are the major disadvantages in the production of molybdenum-99 by this method.
U.S. Pat. No. 5,596,611 discloses a small, dedicated uranyl nitrate (UO2(NO3)2) Aqueous Homogeneous Reactor (AHR) for the production of molybdenum-99 in which the radioactive waste products are recirculated back into the reactor. A portion of the uranyl nitrate solution from the reactor is directly siphoned off and passed through columns of alumina to fix some of the fission products, including molybdenum-99, to the alumina. The molybdenum-99 and some fission products on the alumina column are then removed by elution with a hydroxide and the molybdenum-99 is either precipitated from the resultant solution with alpha-benzoin oxime or purified using other processes. This uranyl nitrate reactor has the advantage of recycling the fission byproducts.
Additionally, the dissolution of uranium or uranium materials to form uranyl nitrate is a common practice in the nuclear industry, as is the use of uranyl nitrate (UO2(NO3)2) in solution reactors. However, the preparation of uranyl nitrate from LEU, including the dissolution of uranium materials in a nitric acid (HNO3) matrix, in the form of a reactor fuel charge for a low power (less than 300 kW) AHR, with specific uranium concentration and pH requirements is a challenge which has yet to be met.
Accordingly, given the above, a need exists in the art for a LEU fuel that is designed to be utilized in an AHR for the production of various medical isotopes and for a corresponding method that produces a LEU fuel suitable for the production of medical isotopes.